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Latest revision as of 23:06, 15 March 2020
| ML19332C852 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 11/17/1989 |
| From: | GULF STATES UTILITIES CO. |
| To: | |
| Shared Package | |
| ML19332C850 | List: |
| References | |
| NUDOCS 8911290080 | |
| Download: ML19332C852 (53) | |
Text
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- DEFINITIONS .
- SECTION
- 1. 0 DEFINITIONS L . .
PAGE L 1.1 ACTI0N...................^...................................., 1-1 1.2 -AVERAGE PLANAR EXP0SURE........................................ 1-1
- 1. 3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..................... 1-1
- 1. 4 CHANNEL CALIBRATION............................................ 1-1
- 1. 5 CHANNEL CHECK........................................,,......... 1-1 -
- 1. 6 CHANNEL FUNCTIONAL TEST........................................ 1-1 1.7 CORE ALTERATION................................................ 1-2 '
K 1. 8 CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY................ 1-2 1.9' CDRE OPEPATING LIMITS REPORP (COLR). . . . . . . . . . . ~. . 1.10 pf CRITICAL POWER RATI0............................. . . . . . . . . . . . . ;. . 1-2 1 11 3.x,00SE EQUIVALENT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '. . . . . . . . . . . 1-2 1.12 p rf DRYWELL INTEGRITY.............................................. 1-2 1.13 3,M l-AVERAGE DISINTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.14 J,.El EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME./........... 1-3 1.15 3 4 4 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE 1-3TINE...... 1.16 3,,3 FRACTION OF LIMITING POWER DENSITY............................. 1-3 1.17 g FRACTION OF RATED THERMAL P0WER................................ 1-3 1.18 g FREQUENCY N0TATION............................................. 1-4 1.19 J< d GASEQUS RADWASTE TREATMENT (OFFGAS) SYSTEM..................... 1-4
- 1. 2o ud IDENTIFIED tEAxAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 .......
1.21 3.e5 ISouTION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.22 J.af LIMITING CONTROL R' 00 PATTERN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.23' ),,4 f LINEAR HEAT GENERATION RATE.................................... 1-4 1.24
/ LOGIC SYSTEM FUNCTIONAL TEST... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1-4 RIVER BEND - UNIT 1 i 8911290080 891117 ,
PDR ADOCK 0500
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u b .: :* ,. . o ..- p .: , l' l- INDEX 1-l L DEFINITIONS l SECTION 1 DEFINITIONS (Continued) PAGE
. 1,25 ~
MEMER(S) 0F THE PUB LIC. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 .... 1.26 I g NINIMUM CRITICAL POWER RATI0..................................... 1-5 1.27 g 0FFSITE 00SE CALCULATION MANUAL.................................. 1-5 1.28 1/ OPERA 8 LE - OPERA 81 LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 ........ 1.29 CPERATIONAL CONDITION - CON 0! TION................................ 1-5 1.30 1 p m SICS TESTS.................................................... 15 1.31 kS6PRESSUREBOUNDARYLEAKAGE........................................ 1-6 1.32 ~
)#1PRIMARYCONTAINMENTINTEGRITY-FUELHAN0 LING..............."..... 1-6
- 1. 3 > pdf'PRIMARYCONTAINMENTINTEGRITY-0PERATING...........'............. 1-6
- 1.34 poiPROCESSCONTROLPR0 GRAM.......................................... 1-6 1.35 3,M RATED THERMAL P0WER.............................................. 1-7 1.36 % REACTOR PRUTECTION SYSTEM RESPONSE TINE.......................... 1-7 1.37 % REPORTABLE EVENT............................................... . 1-7 1.38 3,,41R000ENSITY...................................................... 1-7 1.39 JAI SECONDARY CONTAINMENT INTEGRITY - FUEL BUILDING. . . . . . . . . . . . . .1-7 ....
1.40 34sSECONDARYCONTAINMENTINTEGRITY-0PERATING...................... 1-7 1.41 .pc SNuT00wN mRGI N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-s 1.42 1 / SITE B00MDARY.................................................... 1-8 1.c p1 50 u 0 l n CAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 1.44 y 50uRCe CuECn..................................................... 1-e 1.45 p44 STAGGERED TEST BASI S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1-9 1.46 }AS THERMA L P0WE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1-9 1.47 TINE.............................. 1-9
% TUR8INE BYPASS SYSTEM RESPONSE RIVER BEND - UNIT 1 11 . - - _ . - ~ - - - - _ - , . , , - . - .--,m ~. .,- . . ..__,,. - _ . _ - . _ _ _- . - . - _ . . . . , - , , . . - - , , _ , - . . , . - , . , _ , , - . - - , , , . - - ,
g x . , , . 1 RECElv J
.- SEP 0 S 1988 INDEX SDC LIST OF FIGURES FIGURE TITLE PAGE 3.2.1-1 11 ti. - - Fi = = r iir ei ".;;."...r ^in Rate ( 094)..... ............ .............. 3 -2
- 3. -2 eum Avera anar Linear t' Generation EDELE Tg te (BP85 )......... ................. ...... 3 -3 3.2.1- Maxi Average P1 Linear Heat tion (BP85R82 . . .............. ........... ...... 3/4 2-3.2.1-4 Maximum rage Planar Li Heat'Ge on Rat SRS278).... ............ ............. .. 3/4 2-
.2.1 inum Avera anar Line at Generati Rate (BP 99)....... ............. .......... , 3/4 2-t
- 3) -6 Average ar Linear nerati ate (BP85RB . . ......... .......... .......... /4 2-3.2. - MCPR .............. .......... .......... ....' -9 c 2.0-2 p.............................................. 3/4 2-10 l
3.4.1.1-1 Thermal Power versus Core Flow..................... 3/4 4-3 l 3.4.6.1-1 Minimum Temperature Required Versus Reactor
- Pressure........................................... 3/4 4-24 4.7.4-1 Sample Plan for Snubber Functional Test............ 3/4 7-15 8 3/4 2.3-1 Power Flow Operating Map..................,........ B 3/4 2-6 B 3/4 3-1 Reactor Vessel Water Leve1......................... B 3/4 3-8 8 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) at 1/4 T as a l Function of Service L1fe........................... B 3/4 4-8 l
5.1.1-1 Exclusion Ares..................................... 5-2 5.1.2-1 Low Population Zone................................ 5-3 5.1.3-1 Nap Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents....... 5-4 RIVER BEND - UNIT 1 xxii Amendment No.13,27
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'., ,k- DEFINITIONS SDC
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o:'c m The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential,
.- d overlapping or. total channel steps such that the entire channel is tested.
e.8.- jpCORE-ALTERATION , 4 C-O l
@ j 5 t 1.7. CORE ALTERATION shall be the addition, removal, relocation or movement of I
. e' me fuel sources incore instruments or reactivity controls within the reactor '
T $ r.;.3 press,ure-vesse,l with the vessel head removed and fuel in the vessel. Normal 8 j g ,y movement of the SRMs. IRMs, LPRMs, TIPS or special movable detectors is not ma . o considered a C0RE ALTERATION. Suspension of CORE' ALTERATIONS shall not pre- l u d
- 1 clude completion auf the movement of a component to a safe conservative position.
m en - m l 5-{. g CORE MAXINUN FRACTION OF LINITING POWER DENSITY-am.a gy9S 1.8 g The CORE MAXIMUM FRACTION OF LINITING POWER DENSITY (CMFLPD) shall be the s ow > highest value of the FLPD which exists in the core, l 8 .d % INSERT ,
,@ S 7,$ CRITICAL POWER RATIO '
Om
.S * $ 1. 9 The CRITICAL POWER RATIO (CPR) shall be the ratio of.that power in the
.t ,S g ,o assembly.which is calculated by application of the GEXL correlatico to cause ou e some point in the assembly to experience boiling transition, divided by the 8.1 E E actual assembly operating power.
moo y y;gg DOSE EQUIVALENT I-131 c.o o e - A g d ,. 10 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries e wa s per gram, which alone would produce the same thyroid dose as the quantity and 5 -8 m isottvic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
, o g.t The thyroid dose conversion factors used for this calculation shall be those ae si listed in Table III of TID-14844, " Calculation of Distance Factors for Power
,_,s5',*-landTestReactor51tes."
o ,{ DRWELL INTEGRITY - '
,4 o a
$ ggoce5 1.11 DRWELL INTEGRITY shall exist when: .
2 e $ 8' a. All drywell penetrations required to be closed during accident M $@8e conditions are either: o -mu i5 s8c8 1. Capable of being closed by an OPERABLE drywell automatic isola-y o 'd e - tion system, or m mcem M " 0 o .E 2. Closed by at least one manual valve, blind flange, or deactivated
$ $ y &# automatic valve secured in its closed position, except as pro-pa go ,
c . vided in Specification 3.6.4.
,o gem.c a f ,S .d ' b. All drywell equipment hatches are closed and sealed.
l- I.L c.$. E. e :s
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- c. The drywell airlock is in compliance with the requirements of Specification 3.6.2.3.
8B8% l N$hIVERBEND-UNIT 1 1-2 Amendment No. 29
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$ 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES-(APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits the- ia ci te :: 3. 2.1-1, 2. 2.1-2, 0. 2.1- 2, 3. 2.1-4, 0. 2.1- L , 3. 2.1- 0, 0. 2.1-7 The it:it; ef figure; .2.1-1, 0.2.1-2, 3.2.1-3, 3.2.1--4, N eM ? ?.1-9.*
3.2.1-5, 3.2.1-0, 0.2.1 7 ..d 3.2.1-6 shall be reduced to a.value of 0.84 times the two recir era limit n n operation. shown in the . e LHiiits prcn,'i in the COIR APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTIONI provided in the COLR. With an APLHGR exceeding the limit ef T gere 0. .r l, 3.2.1-2, 3.2.1-3, 3.2.1-4,. 3.2.1-0, 3.2.1-6, 3.2.1-7 or 3.2.1-6, initiate corrective action l within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within i the next 4 hours.
- 2 .
\
l SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determinedyg,Fip;; 2.2.1-1,0.2.1-2,0.2.1-0,0.2.14,0.2.1-5,0.2.1-0,
....,........m
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reacter is operating l with a LIMITING CONTROL R00 PATTERN for APLHGR. )
RECC;yED l di The provisions of Specification 4.0.4 are not applicable. l g 051989 ' 5Dc j
'T;,. 1 ;. ; i.. vi. F ;,. . 3. 2.1-7 .. 4 3. 2.1 0 .. . te :,e used saiy for r;r.;e1 calce!stiaas.
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, I t 0 10 20 30 40 80 CEtVED AvtRA(E'PLANMt DIPOSURE (GWd/t) '
MAR t 41989 FIGURE 3.2.1-7 . I SOC MAXRAAd ANERACE PLANAR IJNEMt MAT (MNDIAll0N RAlt $dAPLMGI) NERSUS AVUtAGE Punan exPosuR - ssans RIVER BEND - UNIT 1 . 3/4 2-68
- Amendment No. 33 l
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-33
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I _ o to no ao 4o -so I RECEI ED 1 RGURE 3.2.1-8 . I NAR 4 1989 WA)0WUW AVERAW PLANAR UNEAR MAT . SDC GENDIA10N RATE OdAPUM) VDtSJS AVUtAM l Puwwt ExP05UK - SS3220 i RIVER BEND - UNIT 1 3/4 2-6C Amendnent No. 33 G
,---w.ee+g.--e w +.,e,.,.-.,,,~.,,,,...,,,-a-. ,,.-,--,,-...,,--,_,em..--.,.._.m.. _,e....m.--..r~m, . . - , .
1,s 3 - Qo t POWER' DISTRIBUTION' LIMITS hwm
'3/4.2.2' APRM SETP0INTS 9
5o LIMITING CON 0! TION FOR OPERATION L
.P 3.2.2. The APRM flow biased simulated thermal power-high scras trip setpoint -
y1 (S) and flow biased neutron flux-upscale control rod block tri go i shall be established'according to fthe4 ollowing relationship /:p setpoint (SR $ e .% . u...s,.a._ . . . . . , Oi ci i Allow alue {[ S 1 (O. + 485)T: _ (0.66W + )T
-S . (0.66W + )T SR8.1 (0 +45%)T
- b. Si e Recircu ton Loop eration Tr Setpoint Allowable alue 1 (0.66W 42.7%)T S1 . 6W + 45. i S., 5 .66W + 36.75 T -- '
_n. - (0.!!" 7 3a. :)T
' S and SoI are-in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation '
flow which produces a rated core flow of 84.5 million 1bs/hr. T = The ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the. CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD). T is applied only if .less' than or equal to 1.0. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to zu of RATED THERMAL POWER. ACTION: With the APRM flow biased simulated thermal power-high seras trip setpoint and/or the flow biased neutron flux-upscale control rod block tri p setpoint less conser- l vative than the value shown in the Allowable Value column' Tor S or S gg, :: Ob: { t' ='=d, initiate correctiv,e action within 15 minutes and adjust 5 and/or S gg to be consistant with the Trip Setpoint value
- within 6 hours or reduce THERMAL POWER to less than 255 of RATED THERMAL POWER within the next 4 hours. l nWith CMFLPD greater than the FRTP, rather than adjusting the APRM.;setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 1005 times CMFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.
RECEIVED RIVER BEND - UNIT 1 3/4 2-7 NOV 301988 AmendmentNo.y SDC
. .. . 3
.. . s .q.
? POWER DISTRIBUTION' LIMITS =. '
3/4.2.3 MINIMUM CRITICAL POWER RATIO-LIMITING CON 0! TION FOR OPERATION
'3.2.3L The MINIMUM CRITICAL POWER RATIO (MCPR) shall be e than both MCPRf and MCPR limits at indicated core flow and TH
- t:-. 1. T' s i.e 3.0.0 " .g .2.0 2.
APPLICABILETY: equal to z!( of RATED THERMAL POWER. OPERATIONAL CONDITION 1, w ' gt{g[:;
. With MCPR less than the appitcable tiCPR Ifnit :h:x h T',...; :.".
; .0.0 :, initiate corrective a: tion t:ithin 15 sitmtes and restore MCPR to 255 of RATED THERMAL POWER within the next 4 hours.' t'wit 1
SURVEILLANCE REQUIREMENTS 4.2.3. a ..__,MCPR a ,_-- shall.,be ___ . determined
. . . __m to be. equal to or greater than the MCPR limit
..............~......-_..
provided in the CDIR. - p
- a. i At least once per 24 hours,
- b. W
. ithin 12 hours after completion of a TWERMAL POWER increase of at least 155'of RATED THERMAL POWER, and '
c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING Col (TROL R00 PATTERN for MCPR. d. The provisions of Specification 4.0.4 are not applicable. RIVER BEND - UNIT 1 3/4 2-8 e e +,w-- w-,,, ,,r.m., ,,....,n.--- . , , _ . . . . - . _ . . . ~ _ . - . . _ _ - ,,-.n
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- CORE FLOW, 5 OF RATED CORE FLOW i
FIGURE 3.2.3-1 WCPR f RhUtBCNO - UMT1 3/42-9 Amendment No. 33 4
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1.0 1M 0 at de 80 N 1m THERMAL POWER,# OF RATED THERMAL PO POURE S.R.8-8 4 RIVER 800 - Ull!T 1 3/4 2-10 i n ____-___-______.,,-__-.,,._.-__.-....-_--mm-m--~ - -, .-
POWER O!STRIBUTION LIMITS RECElVED f.' iE
.3/4.2.4 LINEAR HEAT GENERATION RATE gg g 4 g 5, LIMITING CON 0! TION FOR OPERATION di
**4 g 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not excee _ ^ ' Er/f t fr
- a. O!!"!!:* f;;l ud 13.4 L/fi fei eil ett . f I. ,
APPLICA81LITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25E of RATED THERMAL POWER.
.f i
ACTION: { l With the LHGR of any fuel rod exceeding the lieit, initiate corrective action , within 15 einutes and restore the LHGR to within the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. l SURVEILLANCE REQUIREMENTS 3 4.2.4 LHGR's shall be determined to be equal to or less than the limit: i
- a. At least once per 24 hours, ,
- b. W' thin 12 houts 4fter completion of a THERMAL POWER increase of et east 155 of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING CONTROL R00 PATTERN for LHGR.
, , l J
- d. The provisions of Specification 4.0.4 are not applicable. ,
.o .
TCc m ^G;ese;; N i ;ml e ii r.. ;;;;;; e.-4 00:200. RIVER BEND - UNIT 1 3/4 2-11 Amendment No. 33 0
, . . _ - - . - , . . _ . _ . _ . . . - - - . . . . . . _ . _ . . . - , . . ..__ .- . _ . _ . . . . _ . . _ . - . _ . . - . . . ~ , . , . . . ~ - . - - , . , . - -
.c #
a TABLE 3.3.1 ' REACTOR PROTECTIL-M SYSTEM RESPONSE TIMES E RESPONSE TIME 7 FUNCTIONAL UNIT (Seconds) E 1. Intermediate Range Monftors.
-e a. Neutron Flux - High '
NA 4 9 b. Inoperative NA
- 2. Average Power Range k nitor*:
- a. Neutron Flux - High, Setdown MA
- b. Flew Biased Sieulated Thermal Power - High <0.09**
- c. Neutron Flux - High <0.09
- d. Inoperative NA i 3. Reactor Vessel Steam Ocme Pressurc - High <0.35 i R
- 4. Reactor Vessel Water level - Low Level 3 71.05 i
- 5. Reactor Vessel Water Lesel - High, Leve? 8 71.05 i T
- 6. Main Steam Line Isniation Valve - C %sure 70.09 l 7. Main Steam Line Radiation - High NA
- 8. Drywell Pressure - High NA
- 9. Scram Discharge Volume Water Level - High l a. Level Transmitter NA '
l b. Float Switches MA i
- 10. Turbine Stop Valve - Closure -<0.06 l 11. Turbine Control Valve Fast Closure, W1ve Trip System
! Oil Pressure - Low <0. 0M i 12. Reactor hde Switch Shutdown Position NA i
- 13. Manual Scram NA I
- Neutron detectors are e.weept irem response time testing. Response time shall be measured
- from the detector output or from the input of the first electronic component in the channel.
l **Not including simulated therati power time constant, 5 1 0.0 x x.-C .g
- #Heasured from start of turbine controt valse fast closure. L_ _ g g g g e
I __ .-. _ _ . ~ . . . _ _ _ . _ _ - _
~
is within the limits provided in the Com.j ' TABLE 4.3.1.1-1 (Cu tinued) i M N REACTOR PROTECT 10It 5YSTEM lit $TilWElffAT1011 $URVEILLAIICE REQUIREIEllT5 i es i G i " (f) The LPRMs shall be calileated'at least once per 1000 effective full p hours (EFPII)
- using the TIP systee.
j E j Z (g) Calibrate Rosemount tr4 unit :Walpoint at leest once per 31 days. (h) Verify measured drive flow te be less then er equel to established drive flow at the existing flow control valve position. l (1) This calibration _ , ,____ m shall conshr. f Of verifying the simulated themel peuer time constant to be less (j) ihis' f M on li not requires ta ba OPEP.ASLE when the reacter pressure vessel head is removed per Specification 3.10.1. (k) With any control red withdraun. ht applic41e to control rods removed per Specification 3.9.10.1 or 3.9.10.2. (1) This function is not requiswl to be OPERABLE ene DEVIELL IllTEGRITY is not required per Specifica-tion 3.10.1 i w
- 1 (m) Verify the Tusine THElptAL POWER. typass Valves are closed when T6ElBI4L POWER .is greater then er equel to 4GK RATED i w
{ 4 (n) Theinstruments. CHAINIEL FUIICTICIIAL TEST and MML CALIORATICII shall include the tusine first stage pressure (o) The CHAIINEL CALIBItATI0ll shall esclude the fisw' reference transmitters; these transmitters shell.be i calibrated at least once per 18 months. i (p) This period may be extew$ed to the first refueling estage, not to exceed 9-15-87. l 1
- gO i F s n =a +
- i. a. > . .o , e i$ h 1 g c-On m m o
e p
. -r 2 W,
a.-
" 32 e j Om m
{w .IE ,,, y 5 ,E O- w m * [ c, l ? E O l ==
t ! TABLE 3.3.6-2 " CGisTICL A00 SLOCNTION SE1Polnis , 4
, TRIP FIAICTION TRIP SETPOINfc ALLeimBLE VALUE i
;5i 1. ROD PATTENI CONTROL SYSTih
" a. Law Power Setpelat 27.5 i E of RATED TEWIAL POWER 27.517.5E of RATED T10Est4L
, E POWER l 5 b. High Power 5dpoint i 67.95 of RATED TE WthL POER 168.5 of RATED INEstAL POWER ; As provided in the COIR i ; Flow
.. . Blased .m_
. Neut.res Flur. Hoscale* { ~-
i 0. "" : O i 2. "" : 455*- u . . , . . .h.eudstion j L ; " :-.t h --- 1 0. "" e %.T ' i 3."" ; 39.7E* l b. Insperative NA IIA l c. hale >SE of RATED THENthL POER
- -> E of RATED TE NIAL POWER l d. Ileutron Flum - 71p.,cale
~
Start , $ 1 3 of RATES INEst4L PO K R 1 145 of AATED THEMIAL POWER 1:' 3. SOURCE RAIIGE ImMIT985
- a. Detector not full in IIA IIA Y b. 45 scale < 1 x 10 cps 5
< 1.6 x 105 c,,
C c. Insperative ilA R4
- d. Osunscale > 0.7 cps > 0.5 cys**
- 4. INTENEDIATE RAIIGE IEDNITOR$
c_ m a. Detector not, full la NA IIA l gmO b. 15 scale i 108/125 division of full scale 1 110/125 division of full scale l U) { O y0 c. Insperative tiA IIA
> 3/125 divisier of full O __ < d. Beunscale > 5/125 division of full
- 3m scale scale i
- O 4 5. SCRAft DISCNAPGE VOLT 31E I k= a. Water Level-Migh - Li51i602A < 18.00" < 21.12"
! LI$nSeis -i 18.00" -i 21.60"
- g 1 2 6. REACTOR C00LAIIT SYSTOI REC 1000LATI0ii Flohi
! a. 45 scale i IS E of rated flow $ 111% of rated flew i i
- 1 *The Average Power Eange Monitor red him.k fu6ection is varied as a function of recirculation loop flow (w).
M i The trip setting c! this fonctice e st be maintained in accordance with Specification 3.2.2'q j U **Provided signal to colse ratio is > 2. etherwise setpoint of 3 cps and allowable 1.8 cys. l. k r
> ard limits provided in the GXR. "
4 kv
DESIGN FEATURES
. m =
0 5.3 REACTOR CORE NIAN l SDC ' r= = = : = - l
.1 he reac core s 1 cont n 624 f asse tes. ch ass ly cons ts of zi onium a ey fuel nd water rods ar nged i a nomi 1 8x8 .
arr y. The el rods ntain anium di ide f peller with tive len hs rally nging be en 1 and 150 hes. hose f- asse ies are imited those t t have een anal ed wit RC app ved c es and hods , and ha been s to c ly with 11 of crite a in latest pproved revi on of GE AR (NED 24011-P- -US). k ONTROL ASSE IES > [ 5.3.2 he te or core all e tain 14 cont of crucif rod ass lies, ach con sting ' array stain ss stee tubes urrounde by a er iform aped b' s inless s eel shea . Ea tube s 11 co in 143. inches boron arbide l 4 C X::. , 5.4 REACTOR COOLANT SYSTEN
/ ,
0ESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal tiegradation pursuant to .he a appitcable Surveillance Requirements, a
h, For a pressure ef:
- 1. 1250 psig ca the suctien ofde of the recireuhtion pump. <
- 2. 165C psig from the mircuhtfort pump discharge to tre outlet !
side of the discharge shutaff valve.- p 3. 1550 psig free the discharge shutoff n1ve to the jnt pumps. l
- c. For a temperature of 57 P F l VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 16.000 cubic feet.
l RIVER BEND - UNIT 1 5-5 Amendment No.33 1
l'; c- .L , n-7 ~
- g. , p l
> i i.
-l
. Replacement paragraphs for Technical Specifications 5.3.1, lum ASSDELIES and 5.3.2, CWra0L JOD ASSDeLIES. i MJEL ASSDELIES ,
t 5.3.1' %e reactor-shall contain 624 fuel assemblies. Each assenbly shall consist of.. a matrix of Zircalloy clad fuel rods with an initial camposition of slightly enriched [_ uranium dioxide (002 ) as fuel material. - Fuel assenblies shall be Imited to those ; l- fuel designs apptwed by the NIC Staff for use in WR's. l
- . 6 i
i 0:NTROL RCD ASSDELIES r 5.3.2 2e reactor oore shall contain 145 cruciform shaped control rod assemblies. ; We control material shall be boron carbide powder (B4C ) ' and/or hafniunt netal. %e '
- r. control rod assemblies shall be full length. ,
a t [ t ( U. . h i h I e
. . .v ,-
~
ADMINISTRATIVE CONTROLS SEMIANNUAL EFFLUENT RELEASE REPORT (Continued) - I to MEMERS OF THE PUBLIC due to their activities inside the SITE Si (Figure 5.1.3) during the report period. All assumptions used in making these ' ciuded in these reports. assessments (i.e., specific activity, exposure time and! in accordance with the methodology and p,arameters of the 0FFS ; TION MANUAL (ODCM). The Semiannual Radioactive Effluent Release Report to be submitted 60 days i January 1 of each year shall also include an assessment of radiation doses to the likely nearby most exposed uranium fuel c ME M ER OF THE PUBLIC from reactor releases and other ; and direct radiation)ycle sources (including doses from primary effluent pathways ' for the previous calendar year to show conformance with i 40 CFR Operation. Part 190 Environmental Radiation Protection Standards for N J and gsseous offluents are given in Regulatory Guide 1.109,il 1 The Semiannual Radioactive Effluent Release Reports shall include a list and 1 description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period. { l a The Semiannual Radioactive Effluent Release Reports shall include any changes '
) made= during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the 00CM, as well as a listing of new locations for dose calculations and/or environmental.
cation 3.12.2 monitoring identified by the land use census pursuant to Specifi- ;i gPECIALREPORTS C. 9. 2 Special reports shall be submitted in the following manner: , i J. Spnial reports shall be submitted to the U.S. Nuclear Regulatory jn Commission, Document Control Desk, Washington, DC 205SS, with 6 - i (, copy to the Regional Office of the NRC and a copy to the NRC Resi- , 1 dent Inspector, within the time period specified for each report. i h b. Spacial reports in reperd to C6rbicula will be submitted to the NRC g y ')) within "s0 dsys of identifit.ation of infestation. In setordance with
){s 3
g the settlevent agreement dated October 10, 1984, these reports shall describe the level of. infestation, affected systems and ceasures taken l g to prevent further infestation. l l 6.10 RECORD RETENTION
^ '
6.10.1 L l-Code of Federal Regulations, the following records shall be r least the minimum period indicated. p i 6.10.2 l- The following records shall be retained for at least 5 years: l
- a. 1
[ Records power and logs of unit operation covering time interval at each level. RECEIVED RIVER BEND - UNIT 1 6-19 DEC 151987 ^**"d*'"t * *17 8QC
..--.,,.m..
CORE OPERATING LIMITS REPORT 6.9.3.1 Core operating limits shall be established prior to startup from each reload cycle, or prior to any remaining portion of a reload cycle, for the following: E
- a. The AVERAGE POWER RANGE MONITOR (APRM) setpoints for Specifications 2.2.1, 3.2.2 and 3.3.6.
- b. The AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR) for Specification 3.2.1.
- c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3. ;
! d. The LINEAR HEAT GENERATION RATE (LHGR) of Specification 3.2.4. , e, The REACTOR PROTECTION SYSTEM (RPS) response time for APRM thermal time constant for Specification 3.3.1. , and shall be documented in the CORE OPERATING LIMITS REPORT (COLR). , 6.9.3.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in ti.e NEDE-24011-P-A, " General Electric Sfandard
, Application for Reactor Fuel" (latest approved version).
6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-nydraulic limits, ECCS limits and nuclear transient / accident analysis limits) of the safety enalysis are met. G.9.3.4 The Conf OPERATING LIMITS RCFORT., including any mid eycle revision or supplements shall be providec, upon issuance for each reload cycle, to - the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
' Pcga 13 of 22 i RBS CYCLE 3 COLRi RW. 0 SAMPLE.
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I ? I m u m m m mm m m m m mm m m m m m m m m m m m mmm mm mm m mmmmlemmel 0 10 . 20 30 40 00 AVERAGE PLANA 9 EXPOSURE (GWd/t) l i I FIGURE 2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATIGN MATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE BP85RB163 Ii I L
- \
j
. . . . _ . . . . _ . . _ . - - _ . . .~
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. l L
FIGURE 3 L !- MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLhGR) VERSUS AVERAGE PLANAR EXPOSURE RP85R8248 i' . o
. ~- ,---,.-,,,...,.,-----,,,,v-,-- -.-,r,--,.,ew,w,,,,,,e owe..r---,ve--;c,e-,,-,-w-w, av-,-- ,,v.,+-,- --e ,,- . - , - , , ,y
Paga 15 of 22 l RBS CYCLE 3 COLR'
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FIGURE 4 MAXIMUM AWRAGE PLANAR UNEAR HEAT GENERATION RATE (MAPLHGR) W.RSUS AVERAGE PLANAR EXPOSURE - SPSSRS278 0 ___________.__.s-.____. ,_ _ _ _ _ - . . . .__,_,.,._,,_._.._,,,._,.,_,.,_,ym.,_. _ , . . , , , , , , , , , , _ , , _ , . , . , . _ . _ _ . , , _ . _ _ , _ _ , , ,
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Page 16 or 22 S A M P t. E ass cycts a cota
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- A'4 RACE PLANAR EXPOSUE (GW6/t)
I L . FIGURE 5 MAXIWUW AVERAGE PLANAR UNEAR MCAT GENERATION RATE (MAPLHOR) VERSUS AWGE PLANAR EXPOSURE - BPSSRS299 ,m. _
,- --..w,-,~,. , , , . - , . - - - ~ . . _ - . . . . . . ~ , . . , . . - . . . , , , - - - - , , - - - , . . - - - - , , , . - , - - , , . - , . , , , , , ,-,,-,------...,-,n.
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Page 17 of 22
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. _ . _ . _ _ _ _ . . . , . _ . . _ _ . - - ~ . - _ _ . . _ _ _ , _ . . ' . . . _ - - _ - - - _
S A M P t. E RB C L 3 COLR-n- ECOEECCE
- eMMMMMMMo :
oMMMMMMMMM.
- eMMMMMMMMME%
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l 2 DDDDDDDD ! IIIIIIIl , , 1 3 5 7 91113151719212325272931233537384143454749515355 Fu.1 Typ. 1 A = 383228 E = SP85RB248 8 = 88322C F = SP85RS299
- fl'li';
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, ENCLOSURE II CORE' OPERATING LIMITS REPORT SUBMITTAL I 89-03 !
Page 1 of 22 i RBS CYCLE 3 COLR' l' Rev. 0 SAMPLE , RIVER BEND STATION, OYCLE 3 ! CORE OPERATING LIMITS REPORT (COLR) i I OCTOBER 1989 j l 1 , CONTROLLED COPY 1 I l PREPARED BY DATE: Supervisor, Core Analysis l i River Bend Nuclear Station j l l l l A APPROVED BY: DATE: ' Manager, Engineering River Bend Nuclear Station l 1 APPROVED -BY : DATE: l Facilities Review Committee l River Bend Nuclear Station ; l l APPROVED BY: DATE: Nuclear Review Board ) River Bend Nuclear Station i l l l l l 1
4 . . Page 2 of 22 RBS CYCLE 3 COLR SAMPLE Rev. o LIST OF EFFECTIVE PAGES Page (s) Revision 1-22 0 lr l l l
p ea y+. Page 3 of 22 Ras CYCtE 3 Cota SAMPLE Rev. O INTRODUCTION AND
SUMMARY
o This report provides the values of the Average Power Range Monitor (APRM) scram and rod block setpoints, the 3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) limits, the core flow dependent MINIMUM CRITICAL POWER RATIO (MCPR) limits, MCPR , the thermal power dependent MCPR limits MCPR and the LINEkR HEAT GENERATION RATE (LHGR) limits for River Bend Station, Cycle 3 as required by Technical Specification 6.9.3.1. Per Technical Specifications 6.9.3.2 and 6.9.3.3, these values have been determined using NRC-approved methodology and are established such that all applicable limits of the plant safety analysis are met. f i
w L'. , Page 4 of 22 L SAMPLE RBS CYCLE 3 COLR Rev. 0 l TECilNICAL SPECIFICATION 2.2.1 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The APRM flow biased simulated thermal power-high scram trip setpoints for use in Technical Specification 2.2.1, Table 2.2.1-1 are as follows: ALLOWABLE I FUNCTIONAL UNIT TRIP SETPOINT VALUES
- 2. Average Power Range Monitort
- b. Flow Biased Simulated Thermal Powcr-High '
- 1) Two Recirculation 0.66W+48% 0.66W+51%
Loop Operation
- 2) Single Recircula- 0.66W+42.7% 0.66W+45.7%
tion Loop Operation The relationships for two recirculation loop operation are taken from Reference 4. The relationships for single recirculation loop operation are taken from Reference 5, which is based on Reference 6. l l q l
s= c e oe ,
-c Page 5 of 22 RBS CYCLE 3 COLR Rev. O SAMPLE t
TECHNICAL SPECIFICATION 3.2.1 i POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINFAR HEAT GENERATION RATE The limiting APLHGR value for'the most limiting lattice > (excluding natural uranium) of each fuel type as.a function of AVERAGE PLANAR EXPOSURE is given in Figures 1, 2, 3, 4, 5, 6, 7, and 8. These values were determined with the SAFE /REFLOOD LOCA methodology described in GESTAR-II (Reference 3). Core location by fuel type is provided in Figure 11, which is taken from Reference 3. These figures are used if alternate. calculations are required. The limits of these figures shall be reduced to a value of 0.84 times the two recirculation loop operation limit when in single loop operation (Reference 6). P e I I l l l l 1
f( c i Page 6 of 22 RBS CYCLE 3 COLR SAMPLE Rev. 0 , TECHNICAL SPECIFICATION 3.2.2 POWER DISTRIBUTION LIMITS ' t APRM SETPOINTS ' L The APRM flow biased simulated thermal power-high scram trip setpoint (S) and flow biased neutron flux-upscale R fr use in Technical control rod block trip setpoint-(SSpecification 3.2.2 are as follows:B) , j a) Two Recirculation Loop Operation y ' Trip Setpoint, Allowable Value S (0.66W + 48%)T S (0.66W + Sl%)T , S RB (0.66W + 42%)T S RB (0.66W + 4 5%) T b) Sing)e Recirculation Loop Operation Trip Setpoint Allowable Value S (0.66W + 42.7%)T S (0.66W + 45.7%)T S RB (0. 66W + 36.7%) T S RB ( . 6W + 39.70 T where: S and S RB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 84.5 million lbs/hr. T = The ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD). T is applied only if less than or equal to 1.0. The relationships for two recirculation loop operation are taken from Reference 4. The relationships for single recirculation loop- operation are taken from Reference 5, which is based on Reference 6.
I',' o m e . , 4 e t Page 7 of 22 RBS CYCLE 3 COLR S A ht P L E Rev. O T,ECHNICAL SPECIFICATION 3.2.3 t, POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO The MCPR limits for use in Technical Specification 3.2.3 for MCPR and MCPR are shown in Figures 9 and 10. These values wkre determiEed with the GEMINI methodology and GEXL-PLUS critical power ratio correlation described in GESTAR-II (Reference 1) and are consistent with a Safety Limit MCPR of 1.07. l i l i I
w o , . ,
-t Page 8 of 22 SAMPLE RBS C CLE 3 COLR
!? TECHNICAL SPECIFICATION 3.2.4 POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE The LHGR limits for use in Technical Specification 3.2.4 are 14.4 kw/ft for GE8x8EB fuel and 1.3.4 kw/ft for all other 4 fuel. The GE8x8EB fuel consists of fuel types BS322B and BS322C. Core location by fuel type is provided in Figure 11. The higher limit for GE8X8EB fuel is proprietary to GE and does not appear in Reference 1. The NRC SER on the GE8B design (Reference 2) recognizes the chango to the LHGR limit, ' and the proprietary value is found in References 18 and 19 of Reference 2. k .
r"~", . , 'en g . f, "<?( . pgp};+&I% %; ig g ,g
*- #r' <
(l. . Page 9 of 22 l[ S AL M P L E "88CjCLE3COLR I ' g. ):;; TECHNICAL SPECIFICATION TABLES 3.3.1-2 and 4.3.1.1-1 TheLsimulated thermal power time constant for use in Technical Specification Table 3.3.1-2, Footnote ** ist v.
, 6 + 0.6 seconds, o The maximum simulated thermal power time constant for
~
g use in Technical Specification surveillance Table 4.3.1.1-1 is: i 6.6 seconds
, See Reference 7 for the basis of the time constant.
t
t r , . c .- .\*;
'1
- s o ;c p Page 10-of-22
- S A M P L E. RBS CYCLE 3 COLR t- Rev. 0 TECHNICAL SPECIFICATION 3.3.6 INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATI'N O The APRM flow biased neutron flux-upscale control rod block trip setpoints for use in Technical Specification .
3.3.6, Table 3.3.6-2 are as follows: TRIP FUNCTION l
- 2. APRM
- a. Flow Biased Neutron Flux Upscale *'
TRIP SETPOINT ALLOWABLE' VAL E
- 1. Two Recirculation' O.66W+42%* 0.66W+45%*
Loop' Operation
- 2. One Recirculation 0.66W+36.7%* 0.66W+39.7%*
Loop Operation
- The Average Power Range Mo'nitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of- this' function must be maintained ~in accordance with Specification 3.2.2.
The relationships for two recirculation loop operation ; are taken from Reference 4. The relationships for single recirculation loop operation are taken from Reference 5, which is based on Reference 6. 9
ll
.w#
,=
o !. Page 11 of 22 t SAMPLE RsS CYCLE 3 COLR Rev. O REFERENCES
- 1) NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
- 2) ' Letter, C.O. Thomas to J.S. Charnley, " Acceptance for. 6 Referencing of Licensing Topical Report,"
NEDE-24011-P-A-6, Amendment 10, General Electric Standard Application for Reload Fuel, May 28, 1985. 3)' Document 23A5934, Revision 0, " Supplemental Reload Licensing Submittal for River Bend Station Reload 2, Cycle 3," October'1988.
- 4) GE Design Spec Data Sheet for. Neutron Monitoring System,
-Revision 1, Document Number 22A3739AR, April 1984. ,
t
'5) Engineering Calculation G13.18.6.l*03-0, "APRM Flow Biased ' Scram and Rod Block Setpoints for Single Recirc Loop Operation."-
1 6)- " Single-Loop Operation Analysis for River Bend Station, Unit 1," NEDO-31441, May 1987.
- 7) Later
L: . Page 12 of 22 S A M P L.E ass cycte 3 cota j 13' Rev. 0 gM Ng l t m g 11 ) \ 1
^
k 10 s. k
*.I L % -
%L .
k , s I. N . 7 ,,,mmm ,,mmmmmmm mmmmmmmm mmmmmmmm mmmmmmmmu ; 9 10 . 30 38 de 50 1 AVERAGE PLANAR EXPOSURE (GWd/t) -)
- ~
J l I riOuRE 1 . MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHOR) vtasus AytitAst Puuwt Exposults sPasitsos4 l em en a e e
y - - i2 o 1-4 . , g .a, t ENCLOSURE.III CORE OPERATING LIMITS REPORT SUBMITTAL 89-03 BASES REVISION
SUMMARY
Subject (Section) Revision Limiting Safety System Remove reference to the 6 second simulated Settings (B2.0) thermal time constant and .setpoint formula, i
' Average Planar Linear Heat Provide APLHGR' references to COLR Generator Rate (3/4.2.1) ,
Minimum Critical Power Provide MCPR reference to COLR Ratio (B 3/4.2.3) I i 4 t
A ;(
;1 t 'T 3 POWER DISTRIBUTION-LXMITS 8ASE5 g --
472 MUM CRITICAL 00WER RATIO The'requiredoperatinglimitMCPAs at steady state operating conditions o is
.-+ es !; :ified '- 5;;;ifi;;t.;a 3.2.4 are derived from the established fuel i
. 6perational cladding integrity Safety Limit MCPR cf t.07 and an analysis of abnoratal transients.
For any abnormal operating transient analysis, with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time in Specification during the transient assuming instrument trip settings given. 2.2. L To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO CPR . The t flow, increase in'pressur(e an)d power,ype of transients positive reactivity evaluated insertion, were loss o and coolant provided ' temperature decrease. The limiting transient yields the largest delta MCPR. in the CULR When added to the Safety Limit MCPR ef 1.07, the required einimum operating - limit MCPR e
- Speci*fectie 3.2.2 is obtained-::: i: pr;;; t:d in fip.:r; 3.2.!-1.
Analysisoftransientsoccurringduringsing13recirculationloopoperation indicates that the maximum operating limit MCPR will be bounded by the limits in Sp- !f t::ti-- 3.2.-3 t regions of plant operation,The power-flow map of Figure B 3/4 2.3-1 shows typical o COLR The evaluation of a given tranft ns with the system initial param-eters identified in Reference 2 that are input to a GE core dynamic behavior transient computer program. in Reference 2. The principal The codes used to evaluate transients are described MCPR caused by transient, result of this evaluation is the reduction in provided in the COLR The purpose of the MCPRf and MCPR define operating limits at other than , rated core flow and power conditions. At . lass than 1005 of rated flow and power the required MCPR is the larger value of the MCPR f and MCPR, at the existing core flow and power state. The MCPRy s are established to protect the core from inadvertent core flow increases such that the 99.95 MCPR limit requirement can be assured. The MCPR the correspone,s were calculated such that, for the maximum core the limiting bundle's relative power was adjusted until the MCPR was slightly abcve the Safety Limit. Using this relative bundle power, the MCPRs were calcu-lated at different points along the 10$%-of-rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPRf . RECElVED NOV 301988 RIVER BEND - UNIT 1 8 3/4 2-4 soc Amendment No.12, 31
L Wy - l
.( . -
y Loi LIMITING SAFETY SYSTEM SETTINGS e , BASES- l
?
REACTOR PROTECTI0ft SYSTEM INSTRUMENTATION SETPOINTS (Continued)
, 1 Average Power Ranae Monitor (Continued)
~
, .The'APRM trip system is calibrated using' heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due ;
to transient operation for the case of the Neutron Flux-High setpoint; i.e., 1 for a power increase, the THERMAL POWER of the fuel will he less than that , indicated by -the neutron flux due to the time constants of the heat transfer ' associated with the fuel. For the Flow Biased Simulated Thermal-Power-High
.i' -
setpoint, a time constant Of S ::: 0.0 :e:er.h is introduced into' the flow ' biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as , d r:- h T dl; 2.2.1-1. - "N w nrovided in the CORE OPERATING LIMITS REPORI' ^ (COLR) . ) The APRM setpoints ' fiE'irgiluft'eWp?H'vTdeMiif^foItTi" Safety L Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown. The flow referenced trip setpoint must be adjusted by the specified formula,i- b;if t::ti - 3.2.0 in order to maintain these margins j when CHFLPD is & to FRTP. e 1% plastic strain does not occur in the degraded situation. The scram settTngs and rod block settings are adjusted in accordance with the formula in this specification, when the combination of THERMAL POWER and CMFLPD indicates a peak power distribution, to ensure that L an LHGR transient would not be increased in degraded conditions. RIVER BEND - UNIT 1 B 3/4 2-2 Amendment No. 22, II, 33 l l i' ' s ._ .._. _ . _ _ . _ , _ _ _}}